JPH03105298A - Producing method for solidified body of radioactive waste - Google Patents

Abstract

PURPOSE:To increase an amount of waste which can be filled in a solidification container by using, as a binder, a mixture of one or a plurality of kinds of binder elements which have different distribution factors, and which are selected according to the distribution factors of the binder elements to be appropriately used for each radioactive substances. CONSTITUTION:A concentrated radioactive waste liquid is transfered to a centrifugal thin film dryer 2 from a feed tank 1 to be dried up and pulverized, and then is pelletized by a pelletizer 3 and is filled into a container 4. On the other hand, each binder element is put into a binder tank from tanks 6 and 6 containing different binder elements, along with controlling valves 6' and 6' by a controller 5, based upon a concentration factor of radioactivity concentration and a distribution factor of each binder element. Then, the binder elements are kneaded by a kneader 9 with a predetermined amount of water, and the binder produced in this way, is injected into gaps among the pellets in the container 4 to produce a final solidified body. In this case, a decaying characteristic as a radioactive waste, is improved by around 8 to 10 times.

Description

[Detailed Description of the Invention] [Industrial Field of Application] The present invention is a method for solidifying radioactive waste, which has been reduced in volume, by solidifying it in a container with a solidifying material, and disposing of the solidified radioactive waste in soil etc. as final disposal. This invention relates to a method of guarding a solidified body using a solidifying material with improved radionuclide containment performance, in order to minimize the release of long-half-life radionuclides from the solidified body into the environment through underground water etc. when buried.

[Prior Art] Radioactive concentrated liquid waste and waste resin slurry generated from nuclear power plants and the like have conventionally been solidified as radioactive waste by solidifying them as they are in a container with cement. On the other hand, in recent years, in order to increase the volume reduction rate, concentrated waste liquid and slurry are dried and powdered, or the powder is further granulated into pellets and solidified with cement or other solidifying materials in a container. A method of solidifying it has been used, and recently,
A method is also being developed in which concentrated waste liquid is concentrated into sludge and solidified in a container using a solidifying agent. On the other hand, in Japan, it has been decided that land disposal will be the main final disposal method for solidified radioactive waste, and 19
The plan is being put into practice with the aim of starting the river in 19916. Standards for this purpose are also being developed, one of which is the ``Nuclear Source Materials, Nuclear Fuel Materials, The following Table 1 is listed in Article 13-8 of the Enforcement Order of the Act on Regulation of Nuclear Reactors (Cabinet Order No. 324 of November 21, 1950).

Table 1 This table shows the concentrations of radionuclides in solidified radioactive waste to be disposed of: carbon 4 (denoted as C-14), cobalt 6 0 (Go-60), nickel 6 3 (Ni-
63), stipulates strontium-90 (Sr-90), cesium-137 (Cs-137), and substances that emit alpha rays (hereinafter referred to as alpha waste materials).

[Problem M to be solved by the invention] When solidified radioactive waste is finally disposed of on land, it is important to minimize the leakage of radioactivity into the environment from the solidified solidified material into groundwater in order to reduce the exposure of residents to radiation. This is extremely important from the perspective of preventing pollution and environmental pollution. For this reason, in the design of the final disposal facility, it is planned to apply a so-called human evalia layer using a material such as pennite that adsorbs radioactive materials. However, at the same time, it is desirable to suppress the amount of radioactivity leaching from the solidified radioactive waste to the lowest possible level by increasing the radioactive substance adsorption capacity of the solidifying material itself that constitutes the solidified radioactive waste.

By the way, compared to the conventional radioactive waste solidified body (hereinafter referred to as conventional cement solidified body), which is made by solidifying concentrated radioactive waste liquid or waste resin slurry in a container with cement, the volume reduction rate is even higher. In the above-mentioned solidified material with increased radioactivity, since the amount of radioactivity contained per solidified material is increased, the amount of radioactivity leached from the solidified material tends to increase.

This tendency becomes stronger as the volume reduction rate increases and the amount of radioactivity contained per solidified body increases. Therefore, in order to suppress the amount of radioactivity leached from a solidified material with a high volume reduction rate to the same level or less than the amount of radioactivity leached from a conventional cement solidified material, the higher the volume reduction rate, the higher the radioactive substance adsorption capacity of the solidified material. It is necessary to increase For example, if the volume reduction rate of radioactive waste is twice that of conventional cement 1 solidified material (therefore, the amount of radioactivity contained per solidified material is twice as much), the amount of radioactivity leached under the same conditions will be lower than that of conventional cement. In order to make it equivalent to or lower than that of solidified cement, it is necessary to double or more than double the radioactive substance adsorption capacity of the solidified material.

However, in the past, when selecting a solidifying material for making a solidified body, emphasis was placed on mechanical properties such as strength and fire resistance. Sufficient consideration has not always been given to reducing the amount of leaching.

The purpose of the present invention is to eliminate radiation from the solidified waste by solidifying it using a solidifying material with increased radioactive substance adsorption capacity in producing a solidified radioactive waste with an improved volume reduction rate as described above. The aim is to make the amount of leachate equal to or lower than that of conventional cement solidified bodies.

[Means for Solving the Problems] The present invention proposes a method for producing solidified radioactive waste as set forth in each of the claims.

[Function] In the present invention, in response to the concentrated concentration due to the volume reduction of the radioactive substance according to claim 1 contained in the sterilized radioactive waste described in claim 1, Since the solidifying material components having different distribution coefficients for the radioactive substance are mixed in appropriate proportions and used as the solidifying agent, the amount of the radioactive substance leached from the obtained solidified material is lower than that of conventional cement solidified material. effectively reduced to the same level or below.

[Example] Concentrated radioactive waste liquid as radioactive waste generated from nuclear power plants is dried and powdered for sterilization, then granulated into pellets, filled into a container, and solidified by injecting a solidifying material. Examples of the present invention in this case will be described below. FIG. 1 is a diagram of the process flow, and FIG. 2 is a schematic diagram of the process equipment. The radioactive concentrated waste liquid is sent from a supply tank 1 to a centrifugal thin film dryer 2 where it is dried and powdered, and then pelletized by a granulator 3 and filled into a container 4. On the other hand, from the tanks 6 and 6 containing different solidification material components, the concentration ratio of radioactivity concentration (this depends on the volume reduction ratio by drying and powdering the concentrated waste liquid and pelletizing mentioned above) and the respective solidification materials are obtained. The valves 6' and 6' are controlled by the controller 5 based on a certain distribution coefficient, and the respective solidifying material components are put into the solidifying material tank 7, and together with a predetermined amount of water from the water tank 8, the mixed building Knead in tank 9,
The solidified material thus prepared is injected from the kneading tank 9 into the gaps between the pellets in the container 4 to produce a final solidified material.

The solidified material created in this way contains approximately 8 to 10 times more radioactive materials than a conventional cement solidified material of the same volume made by solidifying the concentrated waste liquid directly in a container with cement. I'm here. That is, the ability to reduce the volume of radioactive waste is improved by about 8 to 10 times. However, on the other hand, the same container contains 8 to 10 times more radioactivity.

Now, Table 2 shows the measured values of the distribution coefficients of each solidifying material component regarding the ions of each radionuclide listed in Table 1.

An example of measuring the distribution coefficient will be explained below. The radioactive concentrated waste liquid is recycled waste liquid of ion exchange resin for desalination at nuclear power plants (
Assuming that the main component is Naz SO (4), a saturated aqueous solution of Na2So was poured into a tank at 50mf,
After adding 0.01 μC i / rn Q of ions of one of the six types of nuclides shown in Table 2, to this solution, one of the solidifying material components shown in Table 2 was cured. Add 1 g of the post-pulverized particles, separate the solution and the solidifying material component after a sufficient time has elapsed to reach adsorption equilibrium, and calculate the 'a degree (μCi/n+Q) of the nuclide in the solution and the solidifying material component. The nuclide concentration (μCi/g) is measured by radiation measurement.

The value obtained by dividing the latter measured value by the former measured value becomes the distribution coefficient of the solidified component with respect to the nuclide.

As shown in Table 2, the distribution coefficient varies greatly depending on the radionuclide and solidification material components. Cement and silicic acid. Particularly, the distribution coefficients of Cs and Sr are significantly different from those of soda.

In the present invention, the amount of radioactivity leached from the solidified radioactive waste to be reduced in volume is made equal to or lower than that from the conventional solidified cement, depending on the radionuclide concentration of the solidified radioactive waste to be reduced in volume. This is to adjust the amount of solidifying material.

Now, let any one of the six types of nuclides shown in Table 2 be the nuclide of interest and represent it by j, any one of the solidification material components shown in Table 2 by k, and the distribution coefficient of k with respect to j. k
It is expressed as dJk.

(1) When using a single solidifying agent component k: (where C,
is the concentration of nuclide j in the solidified body) The target condition is (amount of leachable j from the conventional cement solidified body) ≧ (a) The waste liquid is dried and powdered, or furthermore, the solidified material k is made into pellets. The amount of leaching of j from the volume-reduced solidified material obtained by solidification) is (2). If α is the concentration rate of the concentration of radionuclide j resulting from powdering or pelletizing the concentrated waste liquid, then the above equation (2) becomes K d s x K d jk, that is, ? d, ■ Here, K d J t is the distribution coefficient of cement (represented by k=1).

However, in this case (1) where a single solidifying agent is used, the solidifying agent component used is not a normal cement such as Boltland cement or blast furnace cement (i.e., k≠
1). Generally speaking, the enrichment rate α due to sterilization has almost no dependence on the nuclide; in other words, αj is approximately the same value for all j.

[Example 1] When a dry powder in which Cs is 10 times more concentrated is solidified with sodium silicate, from Table 2, formula (4) is fully satisfied.

In addition, when using a single solidifying material component, according to Table 2, C
Although there is no example of improving the leaching amount of both S and Go compared to conventional cement solidified materials, in reality, C, which is a long half-life nuclide,
It is advisable to pay particular attention to s and try to reduce its elution rate as exemplified in [Example 1] above.

(2) When using a solidifying material prepared by blending a plurality of solidifying material components: In this case, the general formula corresponding to formula (4) is KdJ. Here, K d j * H K d 4 1... are the distribution coefficients of the solidifying agent component a (k=a), the solidifying agent component b (k-=b), etc. used, respectively. ,w. ,W1
... represents the blended weight ratio of each of the solidifying material precepts, W, +W. +...=1...(7)
It is.

[Example 2] If a dry powder with 10 times the concentration of Cs is solidified with a solidifying agent that is a mixture of cement and sodium silicate, what is the equation (6)? dJi (similarly, k = 1 means cement, k = b means sodium silicate), and from Table 2, Kd, ■" l , Kdab =
9 0, so the above formula is l? W+W,=1 Therefore, W■=0.89, W. =0.11, formula (9) becomes 0.89+90X0.11=10.8>1
It becomes 0 and is satisfied.

[Example 3] When dry powder in which Co and Cs are concentrated 10 times is solidified with a solidifying agent that is a mixture of cement, sodium silicate, and oxine-impregnated carbon, equation (6) regarding CO and Cs becomes as follows. .

? d,, (where k=1 is cement, k=b is sodium silicate, k
= c means oxine-impregnated carbon) From Table 2, for Cs, Kd, ■ = 1, Kd3b = 9
0, KdJ− = Kd for IGo, “930
,KdJb=600,Kd,. :27000 Therefore, the following three equations hold true.

? L+Wb+W,=1...(1
3) Solve these three equations and get W■=0.6, Wl,=0.
1. W. =0.3, formula (11), (1
2) is satisfied, and the amount of leaching of both Cs and CO can be reduced compared to conventional cement solidified bodies.

In the above [Example 1], formula (5) is 90 with respect to the target 10, which is too much margin.For example, if the solidifying material is particularly expensive, it is better to use only the necessary amount of solidifying material rather than allowing too much margin. Although it is more costly to use this method, it can be done in Example 2 and Example 3. In the practice of the present invention, the concentration factor α
To actually determine j, sample the concentrated waste liquid from the concentrated waste liquid storage tank or supply tank 1, measure the concentration of solid content (the amount that becomes powder after drying and powdering), and Calculate the concentration α when powdered and further pelletized. As mentioned above, in general, in reality, the enrichment factor α has almost no dependence on the nuclide, and α is practically the same value for all nuclides j. For standard concentrated waste liquid (mainly Na, So. 20wt%), the powderization ratio α=
6 to 8, and in the case of pelletization, α=8 to 10. Note that the nuclide concentration C, can be determined by the gamma ray or beta ray measurement method when performing the sampling measurement described above.

The solidification agent is adjusted each time using the above-mentioned formula based on the concentration ratio α determined by sampling from the waste liquid Ift storage tank or supply tank 1 (or further sampling from the drying powder machine 2). However, in practice, once the volume reduction and solidification treatment system is determined, the concentration ratio α is roughly determined, so it must be adjusted from the beginning accordingly. It is more practical to use a solidified material. For example, in the case of pelletizing, since α is approximately 10, a solidifying material containing sodium silicate as a main component may be prepared in advance and used.

An example of this is the solidifying material (referred to as cement glass) that is a mixture of cement and sodium silicate as described in the previous example.

Basically, the six nuclides shown in Table 2 are selected as the nuclide of interest j, but for operational convenience, the following three nuclides of the nuclides contained in the waste liquid may be used.

More simply, only Cs-137, which has a long half-life (30 years) and is easy to measure because it emits gamma rays, may be the nuclide of interest.

To provide a supplementary explanation on this, in actual operation, it is reasonable to consider not only the concentration rate α but also the nuclide concentration, content, half-life of the nuclide, etc. when determining the solidification material components to be used and their mixing ratio. For example, even if Go-60 (half-life 5.8 years) is mixed in at a concentration 10 times that of Cs-137 (half-life 30 years), the concentration will reach almost the same level in about 20 years. After that, the level of Cs-137 becomes higher, so the management period of the final disposal facility (3 in Japan)
2000). It can be said that it is more rational to select Cs-137 as the nuclide of interest and select the solidifying material.

Figure 3 shows a comparison of the amount of radioactivity released from the solidified bodies. This figure is a diagram in which the Cs release amount of Example I is normalized to the reference II III II and other values are normalized. Example (■) is a case of a solidified product according to an embodiment of the present invention in which the concentrated waste liquid is dried into powder and pelletized and solidified using sodium silicate as a solidifying agent. Example (2) is a case where cement is solidified from the concentrated waste liquid as it is. This is the case with conventional solidified cement that is used as a material and solidified homogeneously. According to the examples of the present invention, it can be seen that the effect of preventing radioactivity from leaching out from the solidified body is superior to that of conventional cement solidified bodies.

In the above explanation, the concentrated waste liquid is dried and powdered.
Or, I have described the case where this is made into pellets and solidified with a solidifying agent. The present invention is applicable not only to this, but also to the case where used ion exchange resin slurry is dried and powdered, or even pelletized and solidified, or when concentrated waste liquid is not powdered, but mud (sludge) is processed. It can be applied to any volume reduction/solidification process, such as when solidifying a substance that has been concentrated to a solid state using a solidifying agent.

[Effects of the Invention] According to the present invention, the radioactivity environment of the nuclide of interest is improved from a solidified body with a higher sterilization ratio than a conventional cement solidified body (therefore, a solidified body with a higher radioactivity concentration in the solidified body). The amount of radioactive waste that can be filled into one solidified container can be increased, and various costs such as waste disposal costs and transportation costs can be reduced. You can save costs.

[Brief explanation of drawings]

FIG. 1 and FIG. 2 are diagrams showing a process flow and an outline of process equipment in an embodiment of the present invention, respectively, and FIG. 3 is a diagram showing a comparison of the amount of radioactivity released. 1... Concentrated waste liquid supply tank 2... Centrifugal filter dryer 3... Granulator 4... Container 5... Controller 6... Solidifying material component tank 7... Solidifying material tank 8...Water tank
9...Mixed tank and others Figure 1 5...i! Il Painting Equipment Guy Kengen F) T Yen

Claims (1)

[Scope of Claims] 1. Radioactive waste obtained by drying and powdering radioactive concentrated waste liquid or waste resin slurry, or by further granulating it into pellets, or by further concentrating the above concentrated waste liquid into slurry. A method for creating a solidified radioactive waste by solidifying radioactive waste in a container using a solidifying material, comprising carbon-14, cobalt-60, cesium-137,
The amount of at least one or more types of radioactive substances contained in the radioactive waste out of the radioactive substance group consisting of substances that emit strontium-90, nickel-63, and alpha rays is leached from the solidified radioactive waste. , one or more types of radioactive substances are added so that the concentration is equal to or lower than that from a radioactive waste solidified body created by solidifying the concentrated waste liquid as it is in a container using cement as a solidifying material. Depending on the radioactivity concentration of and the distribution coefficient of the solidifying material component for each radioactive substance,
A method for producing solidified radioactive waste, characterized in that a mixture of one or more types of solidification material components having different distribution coefficients is used as a solidification material. 2. The method for producing solidified radioactive waste according to claim 1, wherein the solidifying agent component is selected from the group consisting of cement, sodium silicate, zeolite, bentonite, calcium salt, and oxine-impregnated carbon.